Corrosion Science and Technology

  • 제10권4호
  • /
  • Pages.118-124
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  • 2011
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  • 1598-6462(pISSN)
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  • 2288-6524(eISSN)
한국부식방식학회 (The Corrosion Science Society of Korea)
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증기발생기 전열관 재료의 2차측 응력부식균열 민감성

Outer Diameter Stress Corrosion Cracking Susceptibility of Steam Generator Tubing Materials

  • 김동진 (한국원자력연구원 원자력재료개발부) ;
  • 김현욱 (한국원자력연구원 원자력재료개발부) ;
  • 김홍표 (한국원자력연구원 원자력재료개발부)
  • Kim, Dong-Jin (Nuclear Materials Division, Korea Atomic Energy Research Institute) ;
  • Kim, Hyun Wook (Nuclear Materials Division, Korea Atomic Energy Research Institute) ;
  • Kim, Hong Pyo (Nuclear Materials Division, Korea Atomic Energy Research Institute)
  • 투고 : 2011.06.21
  • 심사 : 2011.07.15
  • 발행 : 2011.08.01
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초록

Alloy 600 (Ni 75 wt%, Cr 15 wt%, Fe 10 wt%) as a heat exchanger tube of the steam generator (SG) in nuclear power plants (NPP) has been degraded by various corrosion mechanism during the long-term operation. Especially lead (Pb) is known to be one of the most deleterious species in the secondary system causing outer diameter stress corrosion cracking (ODSCC). Oxide formation and breakdown is requisite for SCC initiation and propagation. Therefore it is expected that a property change of the oxide formed on SG tubing materials by lead addition into a solution is closely related to PbSCC. In the present work, the SCC susceptibility was assessed by using a slow strain rate test (SSRT) in caustic solutions with and without lead for Alloy 600 and Alloy 690 (Ni 60 wt%, Cr 30 wt%, Fe 10 wt%) used as an alternative of Alloy 600 because of outstanding superiority to SCC. The results were discussed in view of the oxide property formed on Alloy 600 and Alloy 690. The oxides formed on Alloy 600 and Alloy 690 in aqueous solutions with and without lead were examined by using a transmission electron microscopy (TEM), equipped with an energy dispersive x-ray spectroscopy (EDXS).

키워드

참고문헌

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